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At least 433 records · Page 24

Preliminary results from Low Pressure Steam Oxidation Testing of ALD ZrN and ZrO2 Coating Deposited over UCN Fuel Kernels

Steam oxidation testing was used to investigate as-developed 100 nm ALD ZrN and 1000 nm ZrO 2 coatings deposited over UC1-xNx fuel kernels. The results of these tests were compared against oxidation of uncoated kernels. This work was performed to support development of coatings which both provide high temperature Zr metal diffusion barrier and also provide resistance against oxidation, especially against high temperature steam and/or air. This work was performed in collaboration with Oak Ridge National Laboratory (ORNL), using ORNL-provided samples. The oxidation tests yielded valuable information, including relative differences in performance of different coating materials and coating thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Binder Jet Printing of H13 Tool Steel

ORNL and Carpenter worked on evaluating H13 powders with different powder size distributions for deposition via binder jet additive manufacturing (AM). Binder jet AM has the potential to be a costeffective technique for fabrication of near net shaped tools with complex geometries and internal features such as cooling channels that can improve the production efficiency of injection/compression molding tools. However, a critical challenge in binder jet AM is the densification of the said tool for it to be usable. Powder size distribution can affect the process settings required to deposit the material and the subsequent sintering behavior. In this collaborative effort ORNL and Carpenter demonstrated that the process variables depend on the powder size distribution while the carbon pickup during sintering depends on the binder burnout atmosphere and temperature. On the other hand, there was no noticeable difference in the microstructural evolution of H13 deposited using the different powder size distributions.

42 ENGINEERING↗

Assembly of MiniFuel Targets for Irradiation of TRISO Fuel Compacts in the High Flux Isotope Reactor

To support the development of Kairos Power’s fluoride-salt-cooled high-temperature reactor, irradiation testing of tristructural isotropic (TRISO) fuel compacts was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental data on TRISO fuel during high particle-power operation and validate fuel performance models. Fuel compacts containing enriched uranium oxycarbide (UCO), natural UCO (NUCO), or uranium dioxide (UO 2 ) TRISO particles were fabricated at ORNL and inserted into MiniFuel targets for HFIR irradiation. Five MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six fuel compacts and will be irradiated in HFIR for four cycles, with target temperatures of 500, 700, and 900°C. This report summarizes the experiment design, test matrix, and fabrication. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Supervisory control and health monitoring framework for large-scale additive manufacturing systems. Final CRADA report

ORNL worked with National Instruments (NI) to develop a large-scale, complex additive manufacturing (AM) systems framework for remote health monitoring and supervisory control. We found the framework, based on the Lincoln Electric Metal AM system located at ORNL’s MDF, capable of controlling the process, leading to improved part quality and digital twin creation.

42 ENGINEERING↗

US Muon Workshop 2021: A road map for a future Muon Facility

The workshop title “US Muon Workshop 2021: A road map for a future Muon Facility” was held virtually on February 1-2, 2021. The workshop aimed to bring together world experts in muon spectroscopy (µSR) and other techniques along with interested stakeholders to evaluate the scientific need to construct a new µSR facility in the United States (US). The more than 200 participants highlighted several key scientific areas for µSR research, including quantum materials, hydrogen chemistry, and battery materials, and how each area could benefit from a new, high flux pulsed muon source. Experts also discussed aspects of the µSR technique, such as low-energy µSR, novel software developments, alongside beam and detector technologies, that could enable revolutionary advances in µSR at a next-generation facility. The workshop concluded with discussion of a concept being developed for a new µSR facility at the Spallation Neutron Source (SNS) of Oak Ridge National Laboratory (ORNL). That novel design concept was first envisioned by many of the same µSR experts at a workshop held previously at ORNL in 2016. The participants expressed that the current design had the potential to be a world-leading µSR facility, and strongly encouraged the principal investigators to continue their work in order to refine the concept and determine instrument parameters that would enable new scientific opportunities.

36 MATERIALS SCIENCE↗

Fracture Toughness Characterization of Generation II FeCrAl Alloys after ~18 dpa Irradiation

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

36 MATERIALS SCIENCE↗

Draft Experimental Plan for Commercial SNF Degradation in Repository Environments with a Focus on Fracture Matrix Degradation

This report documents work performed supporting US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.04, “Inventory and Waste Form Characteristics and Performance.” In particular, this report fulfills the M3 milestone, M3SF-21OR010309072 "Draft Experimental Plan for Commercial SNF Degradation in Repository Environments” as described in work package SF-21OR01030907 "SNF Degradation Testing-ORNL".This experimental plan is a collaboration between ORNL, SNL, ANL, and PNNL. It is intended to support source term model development and spent fuel degradation testing in a range of generic repository conditions. Potential testing and experiments to support validation of the Fuel Matrix Dissolution Model (FMDM) and the overall development/validation of repository source term modeling are discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

TCR Data Management Plan

The Oak Ridge National Laboratory (ORNL) Transformation Challenge Reactor (TCR) program is developing additive manufacturing and artificial intelligence (AI) to deliver enabling technologies for advanced reactors. Through the application of these advanced technologies, the program targets delivering solutions to the high costs and lengthy deployment timelines that threaten the future of nuclear energy—the country’s largest source of carbon-free energy. This document describes the plan for integrating and managing TCR data from multiple sources throughout ORNL and external sources, including novel data sources associated with advanced manufacturing, characterization systems, and the TCR Digital Platform that incorporates various levels of AI and data analytics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Rapid Simulation of Irradiation Damage in PWR Internals (NSUF 2019 Milestone Report)

This document summarizes the progress made by Oak Ridge National Laboratory (ORNL) in collaboration with the University of Michigan on “Rapid Simulation of Irradiation Damage in PWR Internals,” a research project led by the University of Michigan and supported by US Department of Energy Office of Nuclear Energy Nuclear Science User Facilities Consolidated Innovative Nuclear Research program. In this reporting period, samples that were exposed to combined neutron and ion irradiation were prepared and shipped to the University of Michigan for further analysis. This report summarizes the specimen preparation performed at ORNL’s Low Activation Materials Development and Analysis facility. The benchmarking work of microstructural characterization on reactor-irradiated samples with different dose levels was completed.

36 MATERIALS SCIENCE↗

Mineral-Ground Micro-Fibrillated Cellulose Reinforcement for Polymer Compounds

Additive Manufacturing (AM) is quickly becoming a mainstream, energy efficient manufacturing technique for complex and custom components. ORNL has taken polymer AM into the next level with Big Area Additive Manufacturing (BAAM). However, we need to develop the strength of AM materials (specifically polymers) to satisfy the mechanical requirements of structural parts. Also, it is crucial that the developed process can easily be scaled up to have a significant impact in the manufacturing industry of US. Furthermore, renewable, bio-derived AM materials are desirable from a sustainability perspective. In the first phase of this project the mechanical properties of polymer AM components were significantly improved developing a micro fibrillated cellulose (MFC)-reinforced composite material system that is derived from biomass and compatible with BAAM. Unfortunately, the improvements were observed only with freeze-dried MFCs, and freeze-drying is a very slow process and is not industrially scalable. Furthermore, MFCs were not available in large quantities at low cost. However, FiberLean has reached 12,000 tons/year capacity (60,000 tons/year including the minerals). FiberLean offers MFCs at a scale and order of magnitude lower price than current market participants. With their recently developed novel (patent pending) drying process FiberLean can achieve over 50% solid content commercially. In the 2nd phase of the project, the original objective was to investigate this new novel drying process along with compounding with different formulations to optimize and adapt to large scale additive manufacturing. However, the compounded materials prepared via this approach were very brittle and demonstrated poor mechanical performance. Therefore, the project was focused more on a scalable chemical surface modification of MFCs to enable their drying via conventional oven process with minimal agglomeration and better dispersion in PLA polymer matrix. Fiberlean and ORNL have worked together in this project to develop mineral-ground microfibrillated cellulose (MFC) - reinforced additive manufacturing feedstock materials with significantly improved mechanical properties. Properly surface modified and dried Fiberlean cellulose microfibrils led to tensile strength increases up to ~40% and elastic modulus increases up to 70% with the addition of 30% MFC into PLA matrix. The work has led to a high impact journal article and a provisional patent. Opportunities for a follow up work to demonstrate the technology in large scale and application of it to AM applications are currently being searched.

36 MATERIALS SCIENCE↗

Versatile Remediation Module Final Operational Testing Results

Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for performing the on-site remote repair of welded stainless-steel storage containers for spent nuclear fuel and high-level radioactive waste. This report describes the VRM prototype and its design features, components, qualification tests, operational procedures, and demonstration results to support continued long-term storage or off-site transportation of spent nuclear fuel and high-level radioactive waste currently stored in storage containers. A remote (i.e., 100 ft away from the simulated radiative environment) demonstration of the VRM was successfully performed on a full-scale mock-up welded stainless-steel canister. The VRM is designed with features to accommodate remediation techniques beyond those currently selected and described in this report. Therefore, many of the VRM’s features could benefit other remote nuclear or nonnuclear applications. The VRM is envisioned to serve as a development center to facilitate and enhance the further development of new remediation technologies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing

A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. The objective is to explore suitable welding technique and parameters and to determine the helium concentration limitation in terms of irradiated stainless-steel weldability. An additional objective is to develop the knowledge and understanding of cracking mechanisms induced by high concentration of helium in the process of weld repair of the irradiated structural alloys. This report summarizes the experimental welding evaluation on irradiated material from the NRU reactor containing helium concentrations ranging from 12 to 45 atomic parts per million (appm).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High-Accuracy Simulations to Model Pyrometallurgical Processes in a Secondary Lead Reverberatory Furnace

The US manufacturing industry produces about 1.3 million tons of refined lead each year using secondary sources consisting mainly of lead batteries. ORNL is partnering with Gopher resource, the second largest lead recycling company in the United States, and GTI, to develop a high-fidelity CFD model of a directly fired, reverberatory-style, secondary lead furnace. These High Performance Computing (HPC) simulations are aimed to use first principles modeling for combustion and melting processes of the secondary lead feed while accounting for complex interphase interactions between the gas, solid charge (lead) material, slag, and metal phases. Through validation against operating plant data, this effort will enable significant improvements in design, operational parameters, and energy efficiency, thus improving productivity and refractory lifetime of secondary lead melting furnaces. Estimated savings/reduction of, at least, 1 trillion BTU, 1 million ton/year of greenhouse gas emissions, and $\$50$ million/year to the US lead industry can be expected. ORNL resources and expertise in high-performance computing and multicomponent, multiphase flows were utilized to realize this goal while advancing the understanding of the smelting and melting processes occurring within the furnace.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

The First Neptunium Dioxide Neutron Diffraction Experiment at HFIR

Here we present the first neutron diffraction experiment of a 50 mg sample of neptunium(IV) dioxide (NpO 2 ) powder at the High Flux Isotope Reactor (HFIR) using the wide-angle neutron diffractometer (WAND 2 ) HB-2C. We also provide a brief literature survey to pinpoint gaps in the crystal, electronic, and magnetic structure of NpO 2 to determine the optimum experiment to be pursued in the next HFIR user proposal cycle. Characterizing actinide materials with neutron diffraction, inelastic scattering, neutron imaging, polarized neutron studies, and small-angle neutron scattering techniques will increase our understanding of 5f electron behavior and elucidate atomic and magnetic structural properties. This strategy will help fill gaps in the literature and provide theoretical physicists with the data needed to build physical and chemical models to guide and validate computational methods. These methods may be used for predicting structural and chemical features of actinide-based materials. This proof-of-principle effort begins to uncover how to advance actinide science and technology at Oak Ridge National Laboratory (ORNL) using neutron capabilities. This effort has already sparked collaborative discussions between researchers outside the Radioisotope Science and Technology Division, including theoretical physicists from the Nuclear Nonproliferation Division and beamline scientists in the Neutron Scattering Division. We intend to continue pursuing neutron experiments and develop a scientific strategy focused on studying transuranium elements (e.g., Np, Pu, Am, and Cm) using the advanced neutron capabilities at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Application of Shift Ex-Core Calculations to a Detailed MSR Model

This report documents the collaborative research conducted between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) on applying the Monte Carlo neutronics code Shift to calculate the radiation dose rates for a detailed molten salt reactor (MSR) model based on the Molten Salt Breeder Reactor (MSBR). In this work scope, the ANL and ORNL teams worked together and applied Shift to calculate the radiological environments of this detailed MSR facility modeled with explicit geometries. The radiological conditions within the MSR facility were modeled for when the reactor is at two different operation modes: normal full power and drained state. The FW-CADIS hybrid method in Shift was applied successfully to calculate the ex-core neutron and gamma dose rates for the MSBR at full power. Dose rate maps showed that in the current MSBR numerical model, potential pathways exist for neutrons and gammas to stream through the 8-foot concrete shield to reach the top of the MSBR reactor cell. Neutron and gamma dose rates within the drain cell were also calculated for the MSBR at the drained state by integrating the source terms obtained from an ORIGEN-S depletion calculation into the Shift simulation. The results indicate that in the drained state, the delayed gammas from the depleted fuel salt are the main contributors to the dose rates, which suggests that a steel liner in the tank model must be added for calculating the dose rates in the drain cell.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Hospice Landscape Report

In July 2018, CMS requested assistance from Oak Ridge National Laboratory (ORNL) to provide expert data science support aimed at developing algorithms for data mining of medical data for operational and payment purposes. The project is intended to be exploratory: work is aimed at alleviating challenges associated with improper payments, specifically audit methodologies and targeting and changes to risk scores. Project goals include developing new, sophisticated methods for audit targeting and improved profile– payment error correlations, specifically focused on Medicare Part C, the program under which MAOs provide health care services to beneficiaries. ORNL conducted RADV analyses against RAPS and EDS data as well as a hospice landscape analysis per a January 2015 dataset that included Medicare beneficiaries who were in hospice in 2017 and 2018. Ongoing work under this project also involves development of predictive models for RADV investigations and hospice landscape.

97 MATHEMATICS AND COMPUTING↗

COVID-19 Data Curation Effort: An Initial Analysis of the Data

During the COVID-19 pandemic of 2020, major case reporting outlets quickly coalesced around two or three primary vendors. Johns Hopkins University and The New York Times were among the more prominent, and all were of great value to the nation, particularly during the uncertain early stages of the pandemic. They primarily focused on three major attributes: number of new cases, deaths, and recovery. Recognizing that many states were reporting very detailed data sets (e.g., hospital beds) at a county level or finer, the ORNL Pandemic Modeling team embarked on a major data curation effort from March to June 2020 for the purpose of capturing this wealth of detailed data. The challenge of curating this data was daunting. The number of attributes reported by the states grew on almost on a weekly basis. States were routinely shifting their web tool strategies away from easily parsable HTML-based formatting to new Tableau and ArcGIS content. This growth in the sheer number of attributes, combined with the unpredictable shifts in data format, meant an aggressive and agile combination of automated scripting and manual scraping was required to capture new daily streams. Further, the team had to scale up staff and widen its approach for capture and storage. As a result, the team collected more than 11 million data points. Following the close of this data collection effort on June 30 th , 2020, the team embarked on a major effort to appraise what had been collected, including an inventory list, spatial completeness, temporal completeness, scale and geographic characteristics, and a determination. A report on this matter was submitted on September 15 th , 2020, titled “DOE COVID-19 Data Curation Effort: Overview of Data Collection Coverage”. Over 2000 unique attributes had been netted over a wide range of spatial scales, including state, county, zip codes, health regions, and census blocks. Over 11 million individual data points were collected across these attributes, and spatial coverage (in total) included all 50 states and multiple territories. What became apparent in the process is that in the absence of any data standards, many states reported a wide variety of unique attributes that were not always compatible with attributes reported in other states. As time continued, states began adding new attributes and offering finer grain detail in some older attributes. This meant that not all data streams existed for the entire time period; in fact, the number tended to increase dramatically towards the end. Often, states would begin an attribute series and then stop altogether. These highly variable and uncertain conditions illuminated the need for harmonization approaches that would reconcile and conflate changing attribute names and detail over time. For example, grouping racial data reported as either Black or African American, depending on the state, into a single harmonized attribute. These choices would make a within-state analysis possible during the time period and lead to potential between-state analytics later on. This was almost entirely a manual decision process, requiring some subjective decision-making at times, to prevent a fragmented, short-lived collection of time series fragments that would offer few insights into trends, patterns, and correlates. This report imports harmonized data for state and county into the World Spatio-Temporal Analytics and Mapping Project (WSTAMP). WSTAMP is a major space-time analysis and visualization tool developed at ORNL for the National Geospatial-Intelligence Agency specifically for this kind of exploratory analysis. WSTAMP offers a rich analytical and graphical environment consisting of a wide range of analytics. These include time series plots, statistical summaries, data mining techniques, trend and pattern detection, and hypothesis generation.

59 BASIC BIOLOGICAL SCIENCES↗

Material Control & Accountancy for Molten Salt Reactors (FY2021 Report)

There is significant domestic and international interest, investment, and research and development momentum to pursue advanced nuclear reactor technologies. Molten salt reactor (MSR) concepts display the largest variability in fuel type and design features among the current advanced concepts. MSRs have been proposed with various core designs, sizes (power), and fuel cycles. Salt-fueled molten salt systems represent the only advanced reactor type with fuel that is not in a solid form during operation. These “liquid-fueled” MSRs are unique from perspectives of fuel fabrication, spent irradiated fuel and waste components, licensing, and material control and accountability (MC&A) including the potential of fissile material holdup. The liquid fuel salt is the defining distinction in comparison to other advanced reactors that propose TRI-structural ISOtropic particle fuel pebbles, various coolant options (e.g., molten salts or metals, high temperature gas), or small modular alternatives using solid fuel variants including both light water reactors and non-light water reactors. MSRs are appealing to the nuclear energy industry because of the diverse reactor characteristics they can support including various neutron energy spectra, fueling requirements, fuel cycles, and/or fuel utilization. However, because of the significant deviation and diversity of a salt-fueled system compared to traditional solid fuel light water-cooled reactors (LWRs), the history, regulatory licensing framework, modeling capabilities, and supporting engineering technology are either lacking or, in some cases, nonexistent. Therefore, the research community is actively supporting advanced MSR development on many of these fronts in particular to assist MSR vendors with licensing requirements. ORNL is leading the research and development of respective MC&A approaches for salt-fueled MSRs. This report summarizes the research performed at Oak Ridge National Laboratory (ORNL) under the US Department of Energy, Office of Nuclear Energy, Advanced Reactor Safeguards (ARS) program to investigate safeguards and security by design concepts, licensing and regulatory considerations, and dynamic system-level modeling to understand radioisotope concentrations for salt-fueled MSRs. The report builds upon the previous research and literature, identifies the MC&A challenges inherent to a salt-fueled MSR, reviews current regulatory frameworks for LWRs and their applicability towards salt-fueled MSRs, summarizes the status and progress of an MSR dynamic modeling tool, and discusses a prospective MC&A approach based on the Molten Salt Demonstration Reactor (MSDR) model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗